Sharif Digital Repository

Effective delayed neutron fraction βeff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetic parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperatures equal to 20 °C) and hot (fuel, clad and coolant temperatures of 65, 49 and 44 °C, respectively) states using MTR_PC code. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. For the cold state, calculated βeff and Λ by MTR_PC code are 0.008315 and 30.190 μs, respectively. In the hot state, these parameters being... 

In this research, effective delayed neutron fraction (βeff) and neutron generation time (A) of the Tehran Research Reactor (TRR) are calculated for different uranium enrichments from 14.84 w/o to 96.56 w/o U 235 in two states of the TRR, (cold fuel, clad and coolant temperature of 20°C; and hot fuel, clad and coolant temperature of 65, 49 and 44°C, respectively) using the MTR-PC computer code. Comparative analysis shows that both βeff and Λ increase as fuel enrichment decreases. However, variation rate βeff is not the same in two conditions. βeff in the hot state is larger than those calculated in the cold state when fuel enrichment goes more than 83.91%, while the situation is reverse for... 

Various methods have been used for solving the neutron transport equation in the past, and a number of computer codes have been developed based on these solution methods. This paper describes a novel method for the solution of the steady-state and time-dependent neutron transport equation using the duality between neutronic parameters in the method of characteristic (MOC) and the electrical parameters in the cellular neural networks (CNN). The relevant electrical circuit can be simulated by professional electrical circuit simulator software, HSPICE. This software is used for numerical solution of the transport equation only by preparation of appropriate inputs. This method does not need... 

Given the importance and complexity of the three-dimensional (3D) neutron transport equation solution, in the current research, a new Modular Ray Tracing (MRT) Algorithm and 3D characteristic kernel for heterogeneous structures are presented. Improvement of memory management and cache coherency are achieved to some acceptable level. Also, parallel implementation of transport algorithm utilizing OpenMP, cause significant reduction in runtime. To validate our Algorithm, first, a self-constituted pin cell and a lattice arrangement are modeled and results are compared with Monte-Carlo simulation. Second, the well-known 3D benchmark, Takeda model one and two, are investigated and results compared... 

One of the well-known online monitoring techniques used for quality control of bulk samples is Prompt Gamma Neutron Activation Analysis (PGNAA). PGNAA suffers from the so-called matrix effect problems such as density, thickness and moisture content of the sample under study. In this work, an Artificial Neural Network (ANN) model is introduced to deal with these effects. The required spectra for training and testing the proposed ANN model are obtained by Monte Carlo simulation of the gamma-ray spectra recorded in a PGNAA online analyzer system used in cement factories. The gamma-ray spectra related to given set of density, thickness and moisture content are corrected channel-to-channel using... 

This paper presents a comparative study of various soft computing algorithms for reconstruction of neutron noise sources in the nuclear reactor cores. To this end, the computational code for reconstruction of neutron noise source is developed based on the Adaptive Neuro-Fuzzy Inference System (ANFIS), Decision Tree (DT), Radial Basis Function (RBF) and Support Vector Machine (SVM) algorithms. Neutron noise source reconstruction process using the developed computational code consists of three stages of training, testing and validation. The information of neutron noise sources and induced neutron noise distributions are used as output and input data of training stage, respectively. As input... 

This paper presents a developed computational code based on Support Vector Machine (SVM) for reconstruction of energy spectrum of neutron source. To reconstruct unknown energy spectrum using known neutron pulse height distribution, the developed machine is trained by known neutron pulse height distribution of detector and corresponding energy spectrum of neutron source. Validation and testing are the next steps to verify the validity of the calculations done with the developed computational code. The calculated neutron pulse height distributions due to randomly generated energy spectrum using MCNPX-ESUT (MCNPX-Energy engineering of Sharif University of Technology) computational code are used... 

In present study, the neutron noise source is reconstructed using three different unfolding techniques in a typical VVER-1000 reactor core. In first stage, the neutron noise calculation based on Galerkin finite element method (GFEM) is performed; in which the neutron noise in two energy group due to the noise sources of type absorber of variable strength and vibrating absorber is calculated. The neutron noise due to inadvertent loading of a fuel assembly in an improper position (ILFAIP), as a new defined noise source in the neutron noise studies, is calculated as well. In the second stage, the inversion, zoning and scanning methods are applied for reconstruction of the noise source of type... 

In the calculation of the optimum loading pattern for a nuclear reactor, several parameters, including the multiplication factor maximization, the raising of the desired safety threshold during the cycle length and the flattening of the radial power peaking factor should be taken into account. Our evaluation of numerous probable arrangements within the VVER-1000 reactor core (approx. 6 × 10 14 cases) indicated that a direct search of the optimum loading pattern is not feasible. Hence, in this study, the weighting factor method and some constraints on the arrangement of the fuel assemblies were applied so that the above cases were reduced to only 244 possible arrangements. Then, using a... 

In this research, new software package for neutronic calculations, especially kinetic parameters of PWR reactors, has been developed. The program used to link the WIMS-D5, BORGES and CITVAP nuclear codes has been written in Visual C# programming language. This software was used for calculation of kinetic parameters of WER-1000 and NOK Beznau reaction The ration (βeff)i/(βeff)corewhich are an important input data for the reactivity accident analysis, were also calculated. The results were compared with final safety analysis report (FSAR) and published documents. Copyright  

SRAM cache memories suffer from single event upset (SEU) faults induced by energetic particles such as neutron and alpha particles. To protect these caches, designers often use error detection and correction codes, which typically provide single-bit error detection and even correction. However, these codes have low error detection capability or incur significant performance penalties. In this paper, a protected cache scheme based on the variable associativity of sets is presented. In this scheme, cache space is divided into sets of different sizes with variable tag field lengths. The other remained bits of tags are used for protecting the tag using a new protection code. This leads to... 

In this paper, we describe an innovative method to model and solve spatio-temporal behavior of nuclear reactor cores via three-dimensional multilayer cellular neural networks. This method uses electrical elements and the existing duality between neutronic and thermal-hydraulic parameters of nuclear reactors. The relevant electrical circuit can be simulated by existing professional electrical circuit software. This research goes beyond our previous efforts to use a neural computing approach in the nuclear field. Modeling and solving simple nuclear reactor kinetic equations is now expanded to a complete dynamic calculation, integrating the core thermal-hydraulic models and the relevant... 

In an attempt to explore the significance of transport theory in neutron noise, localization of a noise source by Green's function based on transport theory is investigated. There are considerable differences between Green's functions based on diffusion and transport, such as small dimensions, near edges, high heterogeneity medium and high-frequency source of perturbation. These differences are expected to significantly impact unfolding, reconstruction, and identification of the neutron noise source. Improvement in noise source unfolding methods is essential in terms of safety aspects and reactor performance enhancement. Since gaining the ability to monitor nuclear reactor based on noise... 

Proton therapy (PT) is an emerging external beam radiation therapy characterized by superior dose distribution compared to conventional modalities. In the present study, an optimal GATE model was developed and then validated for a double scattering proton nozzle based on the previously constructed model. To this aim, a double scattering treatment nozzle was modeled in the GATE platform. To accelerate the GATE simulations, a virtual range modulation wheel (vRMW) and a variance reduction technique (VRT) were implemented. Proton beam flatness, symmetry, and delivery efficiency, secondary neutron dose, and dosimetric performance were characterized through a set of GATE simulations. The findings...